Contributed by the Pressure Vessels and Piping Division for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received by the PVP Division, February 1, 2000; revised manuscript received April 7, 2000. Technical Editor: S. Y. Zamrik.

Section XI, Subsections IWE and IWL, of the ASME Boiler and Pressure Vessel Code provide requirements to assure that the critical areas of steel and concrete primary containment structures are inspected to detect degradation that could compromise structural and leak-tight integrity. They also specify requirements for preservice examination, acceptance standards, repairs, replacements, and pressure testing. Each nuclear utility in the United States is required to develop and implement a containment inservice inspection program by September 9, 2001, 5 yr after the initial approval in Title 10, Part 50, of the U.S. Code of Federal Regulations. This paper discusses the Code requirements and also provides an insight into the development process and philosophy for containment inservice inspection. [S0094-9930(00)01903-X]

United States Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants.

10.

United States Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.

ACI 201.1 R68, 1984, Guide for Making a Condition Survey of Concrete in Service, American Concrete Institute, (Note: ACI 201.1R-92, issued in March 1997, is a more recent revision than the R68 Revision referenced in Subsection IWL.)